In general, a pressurized water reactor (PWR) in a commercial nuclear plant for use in power generation or hydrogen production needs to satisfy national safety standards, and the minimum number of loops between steam generators and reactor coolant system loops is set to two.
In a typical conventional two-loop PWR, two reactor coolant system loops are disposed symmetrically with respect to a reactor vessel. A reactor coolant pump and a steam generator are disposed in each of the reactor coolant system loops, and the steam generator and the reactor vessel are connected to each other by a hot leg pipe and a cold leg pipe. The reactor coolant pump is disposed on the cold leg pipe. Further, two separate emergency core cooling systems (ECCS) each inject cooling water through an injection nozzle disposed on the cold leg pipe. Two separate ECCS injection pipes are connected to each other by a tie line and are configured to be able to inject water into any of the cold leg pipes.
The reason that two reactor coolant system loops are required in terms of safety as described above is as follows.
(1) Responding to One-Pump Trip Transient
In one-pump trip transient, another reactor coolant pump continues to work to ensure a core flow rate required for cooling core fuel, guaranteeing integrity and reusability of the fuel. If the second reactor coolant pump is not provided, the one-pump trip transient can be a serious accident event entirely equivalent to all-pump trip, making it impossible to satisfy safety standards for a transient event.
(2) Responding to One-Pump Seizer Accident
At the time of a pump seizer accident where a rotor of one reactor coolant pump is suddenly locked during operation, the another reactor coolant pump coasts down according to its inertia to ensure the minimum core flow rate required for cooling a reactor core to thereby reduce the failure of the core fuel to the minimum level. Further, overpressure in a reactor pressure boundary is prevented so as to satisfy safety standards at the accident time. If the second reactor coolant pump is not provided, the core flow rate immediately runs short due to the pump seizer accident, resulting in occurrence of a serious failure of the core fuel and overpressure in the reactor pressure boundary.
(3) Responding to Loss-of-Coolant Accident
If one cold leg pipe is ruptured at the time of loss-of-coolant accident (LOCA), one emergency core cooling system for injecting cooling water into the cold leg pipe may be disabled. Another emergency core cooling system is assumed to be disabled according to a single-failure criterion that one emergency core cooling system is disabled. Even in such a case, the intact emergency core cooling system injects cooling water into the intact cold leg pipe through a tie line so as to cool the core fuel. If the second intact cold leg pipe does not exist, both the two separate emergency core cooling systems may be disabled.
(4) Responding to Steam Generator Tube Rupture Accident
Upon occurrence of a steam generator tube rupture accident (SGTR), the intact steam generator is used to perform primary system depressurization to equalize the pressure of a primary system to the pressure of a secondary system, thereby stopping outflow of nuclear reactor coolant from the ruptured steam generator to the secondary system. If the second steam generator is not provided, depressurization can be achieved only by means of a pilot operated relief valve (PORV) or a pressurizer spray, resulting in a prolonged outflow of the primary coolant to the secondary system. The prolonged outflow of the primary coolant leads to a prolonged discharge of the primary coolant from the relief valve of the steam generator to environment.
For the above reasons, the minimum required number of the reactor coolant system loops of the conventional PWR is set to two. The two-loop PWR is the minimum constituent unit in the conventional PWRs and generally generates a power of 300 MWe to 600 MWe. That is, the two-loop PWR is categorized as a small-sized reactor as commercial reactors. Basically, design concepts of more than 50 years ago are used for the two-loop PWR, and many active components including pumps, etc., are used for a safety system such as the ECCS. Therefore, there was a problem that the safety of a nuclear reactor cannot be maintained when a prolonged station blackout (SBO) occurs.
The number of the reactor coolant system loops has been increased to three or four in order to increase the output power of the PWR. A three-loop PWR generally generates a power of 800 MWe to 900 MWe as a middle-sized reactor. A four-loop PWR generally generates a power of 1100 MWe class or more as a large-sized reactor.
In the conventional four-loop PWR, four reactor coolant system loops are disposed around the rector vessel. As in the two-loop PWR, one reactor coolant pump and one steam generator are disposed in each reactor coolant system loop. In recent years, supersized four-loop reactors capable of generating a power of 1600 MWe class or more are built or planned. The steam generator in such a supersized reactor has a height of as high as about 20 m, and the volume of a containment vessel for housing four such steam generators reaches up to as extremely great as about 80,000 m3. In such a way the reactor output power of a large PWR is increased by simply increasing the number of loops, and its deign is based on the same design concepts as those of more than 50 years ago for the two-loop PWR. Thus, a full passive safety system without any motor-driven pumps has not been adopted.
On the contrary, there is an AP1000 as an example of a passive safety PWR that is capable of increasing the reactor output power without increasing the number of the reactor coolant system loops from two and capable of satisfying the safety standards for accidents with only passive safety systems in which any motor-driven pumps are not used (Refer to, e.g., IAEA-TECDOC-1391, “Status of advanced light water reactor designs 2004”, IAEA, May 2004, p207-p231, p279-p306, the entire content of which is incorporated herein by reference). A description will be given on the AP1000 below with reference to FIGS. 5 to 9.
FIG. 5 is a plan view illustrating a configuration of the reactor coolant system loop (two-loop structure) in a conventional passive safety PWR (AP1000). FIG. 6 is an elevation view illustrating the steam generator and reactor coolant pump of FIG. 5. FIG. 7 is a side view of the steam generator and reactor coolant pump of FIG. 6, which illustrates the inside of the steam generator in a sectional manner. FIG. 8 is an elevation cross-sectional view of a containment vessel used in the passive safety PWR of FIG. 5 and inside thereof. FIG. 9 is a block system diagram of a reactor pressure boundary and a passive cooling and depressurization system (PCDS) used in the passive safety PWR of FIG. 5.
In FIG. 5, a reactor core 1 is housed in a rector vessel 2. Two reactor coolant system loops 50a and 50b are disposed symmetrically with respect to the rector vessel 2. Steam generators 3a and 3b are disposed in their respective reactor coolant system loops. The steam generators 3a, 3b and rector vessel 2 are connected by hot leg pipes 5a, 5b and cold leg pipes 4a, 4b, 4c, 4d. Two reactor coolant pumps 6a and 6b are directly connected to the lower portion of the steam generator 3a, and two reactor coolant pumps 6c and 6d are directly connected to the lower portion of the steam generator 3b. Two separate emergency core cooling systems (ECCS) inject cooling water into the reactor vessel 2 through direct vessel injection nozzles 58a and 58b. Therefore, even if a loss-of-coolant accident in which a cold leg pipe or the like is ruptured occurs, the ECCS is not disabled.
The AP1000 generates a power of about 1117 MWe and thus belongs to four-loop large-sized PWR class in the conventional classification. Thus, four cold leg pipes and four reactor coolant pumps are provided. However, the volume of each steam generator is increased to reduce the numbers of the steam generators and hot leg pipes to two, respectively, thereby succeeding in reducing the number of loops from four to two. This significantly improves the layout efficiency in the containment vessel, thereby succeeding in reducing the volume of the containment vessel to as small as about 58,000 m3. The above advantages are brought about by the improvement of the configuration of the reactor coolant pump directly connected to the steam generator.
FIGS. 6 and 7 are structural views each illustrating an installation method of the steam generator and reactor coolant pump of the AP1000 as a conventional passive safety reactor. The two steam generators have the same configuration, and thus only the steam generator 3a will be described hereinafter.
The two reactor coolant pumps 6a and 6b are directly connected to a channel head 91 disposed at the lower part of the steam generator 3a. 
FIG. 7 illustrates a connecting state between the steam generator 3a and the reactor coolant pump 6a as viewed from the direction perpendicular to the direction of FIG. 6. Since the two reactor coolant pumps are overlapped in this point of view, only the reactor coolant pump 6a is illustrated. Further, the internal structure of the steam generator is also illustrated.
A large number of tubes 92 which are heat exchange pipes having a reverse U-shape are disposed inside a barrel portion 22 of the steam generator 3a. In FIG. 7, only one tube is illustrated. The tube 92 is disposed on a tube sheet 93, and the inside of the steam generator is separated into a primary side 94 and a secondary side 95 by tube sheet 93 and tubes 92.
The inside of the tubes 92 and the space below the tube sheet 93 are referred to as a primary side of the steam generator. The outside of the tubes 92 above the tube sheet 93 is referred to as a secondary side of the steam generator. The primary sides of the steam generators and a system connected to the primary sides of the steam generators are collectively referred to as a primary system. Conversely, the secondary sides of the steam generators and a system (not illustrated) connected to the secondary sides of the steam generators are collectively referred to as a secondary system.
A water plenum 96 which is a primary side component is located below the tube sheet 93. The water plenum 96 is divided into an inlet side and an outlet side by a divider plate 97. An inlet nozzle 98 is located on the inlet side, and the hot leg pipe 5a is connected to the inlet nozzle. The two reactor coolant pumps 6a and 6b are connected to the outlet side of the water plenum. Coolant is sucked by the two reactor coolant pumps and discharged from outlet nozzles 99 of the reactor coolant pumps. The cold leg pipes 4a and 4b are connected to the outlet nozzles 99, respectively. The two reactor coolant pumps are connected to one steam generator, and thus the two cold leg pipes are connected to one steam generator. In FIG. 8, the reactor core 1 is housed inside the reactor vessel 2. The reactor vessel 2 is connected to the two steam generators 3a and 3b by the cold leg pipes 4 (4a, 4b, 4c, 4d) and hot leg pipes 5 (5a, 5b). Further, the reactor coolant pumps 6 (6a, 6b, 6c, 6d) are directly connected to the lower portions of the steam generators 3a and 3b. The components and pipes constituting the reactor pressure boundary are housed inside a containment vessel (CV) 7.
The containment vessel 7 of the AP1000 is the most typical containment vessel, called “large dry CV”, for use in PWRs. The containment vessel 7 is made of steel, because it is designed to be cooled with the external air in case of an accident.
Inside the containment vessel 7, an in-containment refueling water storage tank 8 is disposed. The in-containment refueling water storage tank 8 works as a gravity-driven cooling system (GDCS) if a loss-of-coolant accident (LOCA) in which the cold leg pipe 4 or the like is ruptured occurs. This gravity-driven cooling system cooperating with other passive ECCS submerges the lower part of the containment vessel up to a higher level than the cold leg pipe 4.
After that, it is designed so that a recirculation screen (not illustrated) is opened, introducing water always into the reactor vessel 2 to cool the fuel in the reactor core safely. Once the water introduced into the reactor vessel 2 is heated by the decay heat of the fuel in the reactor core, steam is generated and the steam fills the gas phase of the containment vessel 7, resulting in a rise of the temperature and pressure in the containment vessel 7.
A shield building 71 is built outside the containment vessel 7. A cooling water pool 72 of a passive containment cooling system (PCS) is disposed on the top of the shield building 71. The cooling water pool 72 is filled with PCS pool water 73. In case of a loss-of-coolant accident, the PCS pool water 73 drains onto the containment vessel 7. Air flows into the shield building 71 through an external air inlet 74 and then a natural circulation force raises the air through the gap between an air baffle 75 and the wall of the containment vessel 7 until the air is discharged outside through a heated air discharge 76 formed at the top of the shield building 71. The combination of the drainage of the PCS pool water 73 and the natural convection of air cools the containment vessel 7 safely.
The shield building 71 including its side wall and ceiling portions has a structure endurable against a large plane crash.
In this way, AP1000 can cool the reactor core 1 and containment vessel 7 with an extremely high reliability only by the passive safety systems requiring no external AC power source. However, the plant output power of the AP1000 is as large as 1117 MWe, and the decay heat after an accident is significantly high, so that the PCS pool water 73 depletes in about three days after the accident. Thereafter, the PCS pool water 73 needs to be replenished. That is, the cooling cannot be achieved only by external cooling air.
In FIG. 9, the reactor pressure boundary of the AP1000 is constituted by one rector vessel 2, primary sides of the two steam generators 3a and 3b, two hot leg pipes 5a and 5b connecting therein, four cold leg pipes 4a, 4b, 4c, and 4d, and one pressurizer 80. The cold leg pipes 4a, 4b, 4c, and 4d circulate coolant cooled by the steam generators 3a and 3b into the reactor vessel 2 by means of the driving force of the reactor coolant pumps 6a, 6b, 6c, and 6d. In FIG. 9, only the cold leg pipes 4a and 4c of the four cold leg pipes are illustrated, and reactor coolant pumps 6a and 6b of the four reactor coolant pumps are illustrated. The pressurizer 80 is connected to the hot leg pipe 5a by a surge piping 81.
A passive residual heat removal system 60 (passive RHR) of the AP1000 includes a passive RHR heat exchanger 61. The passive RHR heat exchanger 61 is disposed so as to be submerged in refueling water 66 stored in an in-containment refueling water storage tank 8. The in-containment refueling water storage tank 8 is disposed below an operating deck 90. The passive RHR heat exchanger 61 is connected to the hot leg pipe 5a through a coolant supply piping 62. An inlet valve 63 is disposed in the middle of the coolant supply piping 62. The passive RHR heat exchanger 61 is connected to the cold leg pipe 4a around the outlet of the steam generator 3a through a coolant return piping 65. An outlet valve 64 is disposed in the middle of the coolant return piping 65.
At the normal operation time, the inlet valve 63 is always opened, allowing coolant to be always supplied to the passive RHR heat exchanger 61 through the coolant supply piping 62. Further, at the normal operation time, the outlet valve 64 is always closed.
During the normal operation time of the plant, the outlet valve 64 is closed, preventing the coolant in the passive RHR heat exchanger 61 from passing inside the cold leg pipe 4a for circulation into the reactor vessel 2. However, when water feed to the steam generators 3a and 3b is stopped due to occurrence of a transient such as loss of offsite electric power or loss of feedwater, the outlet valve 64 is automatically opened by a low water-level signal of the steam generators 3a and 3b. As a result, the primary coolant in the passive RHR heat exchanger 61 passes through the coolant return piping 65 and the cold leg pipe 4a to be circulated into the reactor vessel 2. The drive source for the above circulation is the natural circulation force of the primary coolant given by the decay heat generated in the reactor core 1, and an active drive sources such as pumps are not required for the natural circulation in this configuration.
In the case where a steam generator tube rupture accident (SGTR) occurs, the primary coolant outflows from a ruptured location, and the outlet valve 64 of the passive RHR is automatically opened by a low water-level signal of the pressurizer 80. As a result, the primary coolant in the passive RHR heat exchanger 61 passes through the coolant return piping 65 and the cold leg pipe 4a to be circulated into the reactor vessel 2. However, the depressurization of the primary system by the passive RHR is slow and, actually, the emergency core cooling system (ECCS) is automatically activated at the same time to inject cooling water into the reactor vessel 2 for rapid depressurization. Because decay heat removal after the depressurization is performed by the passive RHR smoothly, the ECCS is manually stopped by an operator and the accident is terminated.
Actually, when a steam generator tube rupture accident (SGTR) occurs, a normal operating chemical and volume control system functions to make up for reduction of the water level in the pressurizer 80. This adversely causes a delay of generation of the pressurizer low water-level signal, which may result in an increase in the outflow of the primary coolant to the secondary system. Actually, occurrence of a steam generator tube rupture accident is obvious from rise of secondary system pressure and water level in the ruptured steam generator and thus it is expected that manual depressurization of the primary system can be performed by an operator at an early stage.
However, at this stage, the outflow of the primary coolant is so small that the operator cannot inject ECCS water into the reactor vessel 2 to depressurize the primary system. Further, the depressurization of the primary system using the passive RHR is slow because it only removes the decay heat. Therefore, the operator uses the intact steam generator having a higher cooling and depressurization function to perform the primary system depressurization. The steam generator has heat removal capacity as about 50 times as the decay heat, and thus the primary system depressurization by the intact steam generator is achieved at high-speed. As a result, actually, the accident can be terminated at an earlier stage.
As described above, in the conventional AP1000, it has been necessary to provide two steam generators for terminating the steam generator tube rupture accident at an earlier stage.
In the AP1000, an automatic depressurization system (ADS) is provided for the purpose of achieving the primary system depressurization at high speed upon occurrence of a loss-of-coolant accident (LOCA) and a station blackout (SBO). The automatic depressurization system has four stages: first to third stages 51 to 53 and a fourth stage 68. The first to third stages 51 to 53 are disposed on the pressurizer 80.
The fourth stage 68 of automatic depressurization system is disposed at the same location as the branch position of the coolant supply piping 62 connected to the hot leg pipe 5a. 
Once the automatic depressurization system starts operating, all the stages up to the final fourth stage 68 operate. When the final fourth stage 68 operates, the containment vessel 7 is submerged up to the position of the cold leg pipes 4a and 4b, leading to damage of plant property. As a result the plant cannot be restarted for a long period.
In a steam generator tube rupture accident (SGTR), a damaged location is limited only to the inside of the steam generator although it is an accident. Thus, simply by repairing the tube 92 of the steam generator or replacing it with new one, it is possible to restart the plant in a short period. Therefore, for a steam generator tube rupture accident, it is not allowed to use the automatic pressurization system to depressurize the primary system. It is intended to avoid ADS actuation not only for safety but also property protection both in the primary system depressurization using the intact steam generator by the operator and safety systems of the ECCS and the passive RHR.
Along with global warming and increase in crude oil price, expectations for nuclear power generation plant have increased recently on a global basis. In countries with economic power, construction rush of large nuclear reactors of 1000 MWe class or more is about to start. On the other hand, in developing countries, there is a stronger need for small nuclear reactors of 500 MWe or less in terms of relationship between power demand and the scale of a power network corresponding to the power demand. This trend may increase in the future. However, the small nuclear plants are economically inefficient for their scale disadvantage in terms of unit construction cost. Further, unlike the large nuclear reactors, the small nuclear plants have unique designs so as to make it difficult to prove such unique elemental technologies. Further, siting conditions are often worse than those in the economic powers, so that it is necessary to dispose higher safety than that for the large nuclear plants built in the economic powers. Under the circumstances, demanded is a small PWR capable of increasing economic efficiency by simplification, enhancing safety by a passive safety system, and ensuring reliability by proven elemental technologies common with large nuclear reactors.
The minimum number of loops in the conventional PWR was set to two. However, as the plant output power is increased, the number of loops is increased to three and four, and the structure of the primary system becomes complicated. The AP1000 incorporates simplification by the passive safety system has also a two-loop structure. For further simplification, it is desirable to reduce the number of loops to one. In this case, however, the AP1000 needs to be configured to cope with each of the following events with only one reactor coolant loop: one-reactor coolant pump trip, all-pump trip accident, pump seizer accident, loss-of-coolant accident (LOCA), and steam generator tube rupture accident (SGTR).
Further, in the AP1000, although the containment vessel can passively be cooled by a passive containment cooling system (PCS), it is necessary to replenish cooling water after three days. In the worldwide view, as to the siting conditions of the small nuclear reactors, there exist areas where the small nuclear reactors need to be constructed at sites in the inner portions of a continent and along a river. In the entire operating period, e.g., 60 years, of the plant, shortage of river water can be anticipated to occur. Thus, to cope with such problems with siting conditions, it is necessary to provide a small nuclear reactor provided with a passive containment cooling system capable of ensuring safety of the nuclear reactor without necessity of replenishing cooling water in case of an accident.
Further, more severe natural conditions can be anticipated worldwidely as the siting conditions of the small nuclear reactors. Examples of these include giant cyclones in South-East Asia, the massive earthquake that occurred in Sichuan province of China, and the big Tsunami in the Indian Ocean. Occurrence of a station blackout (SBO) due to a sever natural disaster such as a giant cyclone may prevent a recovery work from being started for a long period of time. The cases of Hurricane Katrina in the United States and the giant cyclone in Myanmar suggest the possibility of such a situation. Similarly, the cases of the massive earthquake in Sichuan province of China and the big Tsunami in the Indian Ocean suggest the possibility of such a situation. Thus, it is necessary to provide a small nuclear reactor capable of performing cooling of the reactor core and containment vessel in a continuous manner even when such a prolonged station blackout occurs. To this end, it is necessary to provide a small nuclear reactor capable of naturally ensuring safety forever without supporting actions such as accident management even if the station blackout continues forever.
In the case of a small nuclear reactor of 500 MWe class or less, thorough simplification needs to be conducted so as to overcome the scale disadvantage. This thorough simplification results in adoption of peculiar and less proven new elemental technologies which is possible only in the individual small nuclear reactor. Most of these new elemental technologies have not been adopted at all and will never be adopted in the future in large nuclear reactors that will surely be constructed. Thus, it is impossible to remove risk of occurrence of defect if a small reactor based on such a new peculiar technology is actually constructed and operated.
An object of the present invention is therefore to use proven elemental technologies and device components of large PWRs or those of large PWRs that will surely be constructed in the future so as to remove the risk associated with new construction and thereby to realize a pressurized water reactor plant which is more reliable, better proved, more simplified, and improved in passive safety.